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Standard Review Plan for Dry Cask Storage Systems (NUREG-1536) |
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Manuscript Completed: January 1997
Manuscript Published: January 1997
Spent Fuel Project Office
Office of Nuclear Material Safety and Safeguards
U.S Nuclear Regulatory Commission
Washington, DC 20555-0001
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The Standard Review Plan (SRP) For Dry Cask Storage Systems provides guidance to the Nuclear Regulatory Commission staff in the Spent Fuel Project Office for performing safety reviews of dry cask storage systems. The SRP is intended to ensure the quality and uniformity of the staff reviews, present a basis for the review scope, and clarification of the regulatory requirements.
Part 72, Subpart B generally specifies the information needed in a license application for the independent storage of spent nuclear fuel and high level radioactive waste. Regulatory Guide 3.61 "Standard Format and Content for a Topical Safety Analysis Report for a Spent Fuel Dry Storage Cask" contains an outline of the specific information required by the staff. The SRP is divided into 14 sections which reflect the standard application format. Regulatory requirements, staff positions, industry codes and standards, acceptance criteria, and other information are discussed.
Comments, errors or omissions, and suggestions for improvement should be sent to the Director, Spent Fuel Project Office, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001.
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Accident-Level. Term used to include both design-basis accidents and design-basis natural phenomenon events and conditions. [See also "Design-Basis ______."] Resistance, response limit, and functional capability requirements apply for conditions and events that exceed "off-normal" or "Design Event II," as described in ANSI/ANS 57.9.
Basic, or fundamental, safety criteria. The following minimal functions for nuclear safety in the design of an ISFSI or MRS facility:
Benchmarking. Validation of the accuracy of a computer code by comparison of the calculated results with those of relevant experiments.
Bias. ANSI/ANS-8.1 defines bias as "a measure of systematic disagreement between the results calculated by a method and experimental data. The uncertainty in the bias is a measure of both the precision of the calculations and the accuracy of he experimental data." See NUREG/CR-6361 for further discussion of bias. Bias defined as the average of the differences between results and measurements is acceptable, provided that one adequately considers the variation in the differences.
Code. Used generically to refer to national or "consensus" codes, standards, and specifications, or specifically to refer to the ASME Boiler and Pressure Vessel Code.
CDE. Committed Dose Equivalent, defined as the total radiation dose equivalent to the body (or specified part of the body) that will be accumulated over 50 years following an intake of radioactive material.
Confinement The spent fuel cladding must be protected during storage against degradation that leads to gross ruptures or the fuel must be otherwise confined such that degradation of the fuel during storage will not pose operational safety problems with respect to its removal from storage.
Containment The assembly of components of the packaging intended to retain the radioactive material during storage.
Confirmatory Calculations. Calculations made by the reviewer to determine whether the package design and specifications meet the regulations. These calculations do not replace the design calculations and are not intended to endorse the applicant's calculations.
Construction. The assembly, fabrication, or putting together of standard parts or components to form structures of systems of a DCSS
Controlled Area. That area immediately surrounding an ISFSI or MRS for which the licensee exercises authority over use and within which ISFSI operations are performed. See 10 CFR 72.3.
Design-Basis _______. The extreme level of an event or condition for which there is a specified resistance, limit of response, and requirement for a given of continuing capability. (Compares with "Design Events III and IV" as described in ANSI 57.9.)
Design Event (I, II, III, or IV). Conditions and events as defined and used for an ISFSI in ANSI/ANS 57.9 (also applicable to an MRS).
Exclusion Area. [Applies to sites with a reactor only] "That area surrounding the reactor, in which the reactor licensee has the authority to determine al activities including exclusion or removal of personnel and property from the area." [10 CFR 100, with additional descriptors included at 10 CFR 100.3.]
Gross Cladding Defect. A known or suspected cladding condition that results in the fuel not meeting its design-basis criteria for dry cask storage. The cask shielding, criticality, thermal, and radiological design analyses typically assume that the cladding provides sufficient structural integrity to retain the fuel pellets in the fuel assembly geometry for normal and accident conditions(1)(2). In addition, both individual fuel rods and fuel assemblies should be intact to preclude fuel handling or operational safety problems during loading and unloading operations. It is the responsibility of the licensee to ensure that fuel placed in dry storage meets the design-basis conditions. This definition is applicable to all phases of dry cask storage (from selection and inspection of the fuel before loading until the fuel is unloaded from the cask or the cask is placed in a permanent repository). Alternative means, such as canning, will be required for dry cask storage of fuel that does not meet design-basis conditions.
Hard Receiving Surface for a horizontal or vertical drop need not be an unyielding surface; rather the receiving surface may be modeled as a reinforced concrete pad on engineered fill.
Important Confinement Features. Term used in ANSI/ANS 57.9, but not acceptable to the NRC. (Per RG 3.60, "important to safety" should be substituted for "important confinement features" in the standard.)
Important to Safety [also "Important to Nuclear Safety"]. A function or condition required to store spent fuel of high-level waste safely. To prevent damage to the spent fuel or the high-level waste container during handling and storage, to provide reasonable assurance that spent fuel or high level radioactive waste can be received, handled, packaged, stored, and retrieved without undue risk to the health and safety of the public.
Independent Calculation Calculations separate from the applicant's. Input data should be taken from primary sources such as the package drawings and manufacturer's specifications. Models should be developed separately by the reviewer. To the extent possible, different techniques, codes, and cross section sets or other derived data sets should be used.
Intact Cladding. Spent fuel cladding that does not have gross cladding defects (see Gross Cladding Defects).
Mixed waste. Waste material that is hazardous because it contains both radioactive material as well as chemical, toxic, incendiary, or other hazards.
MofS. Margin of safety, which may be defined as identical to factor of safety, f.s. = capacity/demand (with minimum acceptable MofS 1.0), or as a true margin, where MofS = f.s.-1 = (capacity/demand) - 1 (with minimum acceptable MofS 0.0.
NDE: Nondestructive examination: testing, examination, and/or inspection of a component which does not affect the use of the component. NDE can be broadly divided into three categories: visual, surface, and volumetric examinations. [Additional information may be found in the ASME B&PV Code, Section V, Nondestructive Examination, Appendix A.]
NDE related terms in order of increasing severity:
discontinuity: an interruption in the normal physical structure of a material. Discontinuities may be unintentional, such as those formed inadvertently during the fabrication process, or intentional, such as a drilled hole.
indication: detection of any discontinuity using an NDE method.
flaw: detection of an imperfection or unintentional discontinuity using an NDE method.
defect: a flaw which, due to its size, shape, orientation, location, or other properties, is rejectable to the applicable construction code. Defects may be detrimental to the intended service of a component and the component must be repaired or replaced.
Common NDE examination methods include:
LT leak testing
MT magnetic particle examination
PT liquid penetrant examination
RT radiographic examination
UT ultrasonic examination
VT visual examination
destructive
examination: testing, examination, and/or inspection of a component which results in the destruction of the component.
Normal. The maximum level of an event or condition expected to routinely occur. The ISFSI or MRS is expected remain fully functional and to experience no temporary or permanent degradation from normal operations, events, and conditions. (Compares to "Design Event I" of ANSI/ANS 57.9.) Events and conditions that exceed "normal" levels are considered to be, and to have the response allowed for, "off-normal" or "accident-level" events and conditions.
Off-Normal. The maximum level of an event or condition that although not occurring regularly can be expected to occur with moderate frequency and for which there is a corresponding maximum specified resistance, limit of response, or requirement for a given level of continuing capability. (Similar to "Design Event II" of ANSI/ANS 57.9.) ISFSI SSC are expected to experience off-normal events and conditions without permanent deformation or degradation of capability to perform their full function (although operations may be suspended or curtailed during off-normal conditions) over the full license period.
Other radioactive wastes. Components generally associated with the spent fuel, e.g. Control Assemblies (Rods) BWR fuel channels etc.
Quality Group. NRC classification of SSCs by degree of importance to nuclear safety (NUREG-0800, §3.2.2, and Regulatory Guide 1.26) for reactor systems and adapted to use with ISFSI as follows:
Radwaste. Waste that is hazardous because it contains nuclear materials (may be high- or low-level.
Ready Retrievability. Capability to return the stored radioactive material to a safe condition without the release of radioactive materials to the environment or radiation exposures in excess of the limits defined by 10 CFR 20 [10 CFR 72.122(h)(5)]. ISFSI and MRS storage systems must be designed to allow ready retrieval of the stored spent fuel or high-level waste (MRS only) for compliance with 10 CFR 72.122(l).
Restricted Area. "Any area to which the licensee controls access to protect individuals from exposure to radiation and radioactive materials." [10 CFR 20]
Safety Analysis Report. In the context of the FSRP, the report submitted by the license applicant in compliance with 10 CFR 72, Subpart B or I. The fundamental contents of the report are described at 10 CFR 72.24. Guidance regarding the content of the report is provided by Reg. Guides 3.48, 3.61 and 3.62. For the staff review, the SAR is considered to constitute the actual SAR submitted with the application, along with supplemental data submitted with the application and supplemental data and responses submitted following the application during the NRC staff review and evaluation. The effective SAR is considered by the staff to be that submitted, as amplified and/or modified by the supplemental and later submissions that are docketed.
Safety Evaluation Report. In the context of the FSRP, the report prepared by the NRC staff to present findings and recommendations relating to the acceptability of the applicant's safety analysis and other required submissions. The SER also identifies the bases for those recommendations and the recommended technical specifications ("operating controls and limits" or "conditions of use").
Unrestricted Area. "Any area to which the licensee need not control access in order to protect individuals from exposure to radiation and radioactive materials." [10 CFR 20]
Volume %. The percent of a mole of the material that is present in a volume equal to the standard volume for the material as a gas; the volume occupied by one mole of the material as a gas at standard conditions for gases (760 mm Hg (760 torr) pressure and 0C temperature).
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This standard review plan (SRP) provides guidance for use by staff reviewers from the U.S. Nuclear Regulatory Commission (NRC), Office of Nuclear Material Safety and Safeguards (NMSS), Spent Fuel Project Office, in performing safety reviews of applications for approval of spent fuel dry cask storage systems (DCSS). The principal purposes of the DCSS SRP are to ensure the quality and consistency of staff reviews and to establish a well-defined basis from which to evaluate proposed changes in the scope of reviews.
Other purposes of this SRP are to ensure wide availability of information about regulatory matters, to improve communication, and to help interested persons and the nuclear power industry better understand the staff review process.
The regulations (10 CFR Part 72) that govern the storage of spent nuclear fuel are largely performance based. An example of a performance based regulation can be found in 72.122 Overall requirements:
(a) Quality Standards. Structures, systems, and components important to safety must be designed, fabricated, erected, and tested to quality standards commensurate with the importance to safety of the function to be performed.
This SRP describes the process and provides the reference documents reviewers need to evaluate what "commensurate with importance to safety" means and to evaluate it constantly with respect to the many different designs for DCSS that may be submitted for approval.
A DCSS may be used to store spent nuclear fuel under either a site-specific or general license to operate an independent spent fuel storage installation (ISFSI). At present, any holder of an active reactor operating license under Title 10, Part 50, of the U.S. Code of Federal Regulations (10 CFR Part 50), has the authority to construct and operate an ISFSI under the provisions of the general license. Requirements for construction and pre-operational activities of such an ISFSI are discussed in Subparts K and L of 10 CFR Part 72. The requirements for pursuing a site-specific ISFSI license are discussed in Subparts B and C of 10 CFR Part 72. Regardless of the license type, the NRC staff must review and approve the cask design that will be used in an ISFSI before spent fuel loading begins. This SRP describes the methods used by the NRC staff to conduct such a review.
The DCSS safety review is primarily based on the information provided by an applicant, or cask vendor, in a safety analysis report (SAR). Sections 72.24 and 72.230 of 10 CFR Part 72 require inclusion of an SAR in each application for a license to store spent nuclear fuel or for approval of spent fuel casks. Before submitting an SAR, an applicant should have designed and analyzed the storage cask system in sufficient detail to conclude that it can be properly fabricated and safely operated without endangering the health and safety of the public. The SAR is the principal document in which the applicant provides the information that reviewers need in order to understand the bases for reaching the conclusion that the storage cask is acceptable for use.
Section 72.24 specifies, in general terms, the information to be supplied in an SAR. The specific information required by the staff for evaluation of an application is identified in Regulatory Guide (RG) 3.61, "Standard Format and Content of Topical Safety Analysis Reports for a Spent Fuel Dry Storage Facility." The sections of this SRP are keyed to the standard format defined in RG 3.61. Similar information is also provided in RG 3.62, "Standard Format and Content for the Safety Analyses Report for On-Site Storage of Spent Fuel Storage Casks."
This SRP is written to address a variety of site conditions and cask system designs. Each section presents the complete review procedure and all current acceptance criteria for all pertinent areas of review. However, for any given application, the staff reviewers may select and emphasize the particular aspects of each SRP section that are appropriate for a given application. In some cases, a cask feature may be sufficiently similar to that of a previous cask so that a de novo review of the feature is not needed. For these and other similar reasons, the staff may not carry out in detail all of the review steps listed in each SRP section in the review of every application. Conversely, the staff may find it necessary to ask additional questions or probe areas in greater depth, in order to adequately review a particular design. Review plans have not been included for SAR sections that consist of background or design data that are included for information or for use in reviewing other SAR sections.
The individual SRP sections address, in detail, the matters that are reviewed, the basis for the review, how the review is accomplished, and the conclusions that are sought. Each SRP section is organized into seven subsections, as follows:
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This subsection states the purpose and scope of the review.
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This subsection describes the systems, components, analyses, data, or other information that are reviewed as part of the given SAR section. It also discusses the information needed or coordination expected from reviewers of other SAR sections in order to complete the subject technical review.
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This subsection summarizes the applicable sections of 10 CFR Part 72 pertaining to the given SAR section. This list is not all inclusive (e.g., some parts of the regulations, such as 10 CFR Part 20, are assumed to apply to all sections of the SAR).
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This subsection addresses the design criteria and in some cases specific analytical methods that NRC staff reviewers have found to be acceptable for meeting the regulatory requirements, specified in 10 CFR Part 72, that apply to the given SAR section.
These acceptance criteria typically set forth the solutions and approaches that staff reviewers have previously determined to be acceptable in dealing with a specific safety problem or design area that is important to safety. These solutions and approaches are discussed in the SRP so that staff reviewers can take uniform and well-understood positions as similar safety issues arise in future cases. Like regulatory guides, these solutions and approaches are acceptable to the staff, but they are not the only possible solutions and approaches. Applicants should recognize that, as in the case of regulatory guides, substantial staff time and effort has gone into developing these acceptance criteria, and a corresponding amount of time and effort may be required to review and accept new or different solutions and approaches. Thus, applicants proposing solutions and approaches to new safety issues or analytical techniques other than those described in the SRP should expect longer review times and more extensive questioning in these areas. An alternative is to propose new methods on a generic basis, apart from a specific license application. Such an alternative proposal could consist of a submittal of a Topical Safety Analysis Report (TSAR). This type of application could form the basis for either a change in the staff interpretation of the regulatory requirements or support a request for rulemaking to change the requirements themselves.
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This subsection discusses how the review is to be accomplished, including the general procedure that reviewers follow to establish reasonable verification that the applicable safety criteria have been met.
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This subsection presents the type of conclusion that is sought for the given review area. For each area, a conclusion of this type is included in the safety evaluation report (SER) in which the staff reviewers publish their findings. The SER also describes which aspects of the review were selected or emphasized; which matters were modified by the applicant, require additional information, will be resolved in the future, or remain unresolved; where the cask's design deviates from the criteria stated in the SRP; and the bases for any deviations from the SRP.
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This subsection lists the references commonly used in the review process for the given subject area.
The SRP and RG 3.61 are directed toward storage cask systems designed for spent fuel with zircalloy cladding. Staff reviewers may adapt the SRP as needed for use in reviewing other storage designs and spent fuel types.
The SRP results from years of staff experience establishing and using regulatory requirements to review SARs and to evaluate the safety of spent fuel storage system designs. This SRP may be considered a part of the continuing regulatory standards development process and documents current review methods.
The SRP may be revised and updated as the need arises to clarify the content, correct errors, or incorporate modifications approved by the Director of the SFPO. Comments, suggestions for improvement, and notices of errors or omissions will be considered by and should be sent to the Director, Spent Fuel Project Office, Office of Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory Commission, Washington, DC 20555.
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The purpose of reviewing the general description of the cask or dry cask storage system (DCSS) is to ensure that the applicant has provided a non-proprietary description that is adequate to familiarize reviewers and other interested parties with the pertinent features of the system.
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The general description should enable all reviewers, regardless of their specific review assignments, to obtain a basic understanding of the DCSS, its components, and the protections afforded for the health and safety of the public. Regulatory Guide (RG) 3.61(3) provides general guidance regarding information that should be included in the general description. Because much of the information relevant to this initial aspect of the DCSS review is presented in more detail in other chapters of this standard review plan (SRP), this chapter focuses on familiarization with the DCSS and should be consistent with the remaining sections of the safety analysis report (SAR). Specifically, this focus may encompass the following areas of review:
1. DCSS description and operational features
2. drawings
3. DCSS contents
4. qualifications of the applicant
5. quality assurance
6. consideration of 10 CFR Part 71(4) requirements regarding transportation
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In general, this initial aspect of the DCSS review seeks to ensure that the applicant's general description of the DCSS fulfills the following acceptance criteria:
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Sketches and diagrams, if presented in this section, should be compared with the detailed drawings presented elsewhere in the SAR. If the application includes proprietary drawings and descriptions, that will remain proprietary upon approval of the license or certificate, the sketches, drawings and diagrams that provide the general description and operational features need not show the proprietary features. This may be achieved by depicting less detail or by illustrating generic components which fulfill the design function that differ from the actual design. However, these representations should show the operational concept and safety-related features in sufficient detail to form an acceptable basis for public review and comment, as necessary for public hearings.
In addition to information on a single DCSS, the application should describe any limitations on the arrangement of DCSS arrays. For some DCSS, this may be minimum spacing between DCSS, maximum density of DCSS in an array, and/or total number of DCSS or amount of spent fuel that may be stored at a single independent spent fuel storage installation (ISFSI). The acceptable limitations should be included among the conditions for use in the SER (see Chapter 12 of this SRP). However, for DCSS systems such as those with metal confinement vessels stored in a concrete vault, information on the configuration of vault compartments and horizontal/vertical arrangement is necessary.
The level of detail needed in the drawings is generally assessed by each reviewer during the evaluation of specific sections of the SAR. Particular attention should be devoted to ensuring that dimensions, materials, and other details on the drawings are consistent with those described in both the text of the SAR and those used in supplementary analysis. If size reduction has rendered any information unclear or illegible, reviewers should request that the applicant provide larger or full-size drawings.
Drawings applicable to the SAR review should be identified by number and revision in Section 12 of the SER.
If the applicant proposes the storage of spent fuel with gross cladding defects or storage of non-fuel core components that do not have an integral confinement boundary, the range of permissible conditions for the stored material must be defined. If the DCSS system is intended to be used to store fuel with gross cladding defects or an integral confinement boundary when placed in the confinement DCSS, the possible range of conditions of the fuel or components should be stated. 10 CFR 72.122(h)(1) requires "canning" or use of other acceptable means for storing fuel with cladding that is not or may not remain intact and for unconsolidated assemblies (without intact cladding). The application, therefore, should address the following basic requirements:
If the requested approval is to address the possible use of the DCSS system for storing non-fuel core components, the application should present summary descriptions of those components. Also, if the components are degraded (e.g., the component does not provide adequate confinement under design basis conditions to contain radioactive gas or other dispersable radioactive materials), the application should describe the possible conditions and alternative confinement methods, if any.
Because applications for certification under 10 CFR Parts 71 and 72 are sometimes submitted concurrently, the final (approved) version of such documents may not be available at the time of initial DCSS SAR submission. Nonetheless, applicable documentation of the Part 71 certification, including questions and responses from the related review, is generally provided to the Part 72 review team, as appropriate. Substantial coordination of the Part 71 and Part 72 reviews is necessary to ensure consistency and avoid duplication of effort. The applicant should have a process for promptly informing each of the review teams about DCSS system design changes precipitated by any concurrent safety reviews. Provisions for communicating these changes should be addressed by and discussed with the applicant.
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Review the 10 CFR Part 72 acceptance criteria and provide a summary statement for each. These statements should be similar to the following model:
A general description and discussion of the DCSS is presented in Section(s) _____ of the SAR, with special attention to design and operating characteristics, unusual or novel design features, and principal safety considerations.
Drawings for structures, systems, and components (SSCs) important to safety are presented in Section _____ of the SAR. A listing of those drawings that were relied upon as a basis for approval appears in Section 12 of the Safety Evaluation Report (SER).
Specifications for the spent fuel to be stored in the DCSS are provided in SAR Section _____. Additional details concerning these specifications are presented in Chapter 2 of both the SAR and SER.
The technical qualifications of the applicant to engage in the proposed activities are identified in Section _____ of the SAR.
The quality assurance program, and implementing procedures, are described in Section 13 of the SAR.
The [DCSS system designation] [has been/is/is not being] certified under 10 CFR Part 71 for use in transportation. A copy of the SAR and Certificate of Compliance issued under 10 CFR Part 71 are on file with the NRC under Docket No. ________ [if applicable].
The staff concludes that the information presented in this section of the SAR satisfies the requirements for the general description under 10 CFR Part 72. This finding is reached on the basis of a review that considered the regulation itself, Regulatory Guide 3.61, and accepted practices.
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The purpose of evaluating the principal design criteria related to structures, systems, and components (SSC) important to safety is to ensure that they comply with the relevant general criteria established in 10 CFR Part 72(6), further guidance can be found in NUREG/CR-6407(7) "Classification of Transportation Packaging and Dry Spent Fuel Storage System Components According to Importance to Safety." Material provided in this chapter will form the basis for accepting the safety analysis report (SAR) for staff review.
The applicant should present details of the principal design criteria in either Section 2 or defer the details to the associated sections of the SAR. If the applicant chooses deferral, a general reference to these criteria must be presented. Regulatory Guide (RG) 3.61(8) provides general guidance concerning information that should be included in the principal design criteria for a dry cask storage system (DCSS). In general, these criteria include specifications regarding the fuel or other material to be stored in the DCSS, as well as the external conditions that may exist in the casks operating environment during normal and off-normal operations, accident conditions, and natural phenomena events. A detailed evaluation of how the DCSS design meets the principal design criteria should be presented in Sections 3 through 14 of the safety evaluation report (SER).
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The following areas of review have been adopted by the NRC staff, and include those areas noted in RG 3.61:
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The applicant must specify the design bases and criteria all SSC that are important to safety. [10 CFR 72.24(c)(1), 72.24(c)(2), 72.120(a), and 72.236(b)]
SSC important to safety must be designed, fabricated, erected, and tested to quality standards commensurate with the importance to safety of the function to be performed. [10 CFR 72.122(a)]
The applicant must identify all codes and standards applicable to the SSC. [10 CFR 72.24(c)(4)]
The design-basis earthquake must be equivalent to or exceed the safe shutdown earthquake of a nuclear plant at sites evaluated under 10 CFR Part 100(10). [10 CFR 72.102(f)]
The DCSS must maintain confinement of radioactive material within the limits of 10 CFR Part 72 and Part 20, under normal, off-normal, and credible accident conditions. [10 CFR 72.236(l)]
The DCSS must be designed and fabricated so that the spent fuel is maintained in a subcritical condition all under all credible normal, off-normal, and accident conditions. [10 CFR 72.124(a) and 72.236(c)]
The spent fuel cladding must be protected during storage against degradation that leads to gross ruptures, or the fuel must be otherwise confined such that degradation of the fuel during storage will not pose operational safety problems with respect to its removal from storage. [10 CFR 72.122(h)(1)]
Storage systems must be designed to allow ready retrieval of spent fuel waste for further processing or disposal. [10 CFR 72.122(l)]
The DCSS must be designed to provide adequate heat removal capacity without active cooling systems. [10 CFR 72.236(f)]
During normal operations and other anticipated occurrences, the annual dose equivalent to any real individual who is located beyond the controlled area must not exceed 25 mrem to the whole body, 75 mrem to the thyroid, and 25 mrem to any other organ as a result of exposure to (1) planned discharges to the general environment of radioactive materials except radon and its decay products, (2) direct radiation from operations of the ISFSI or monitored retrievable storage (MRS), and (3) any other radiation from uranium fuel cycle operations within the region. [10 CFR 72.24(d), 72.104(a), and 72.236(d)]
Any individual located at or beyond the nearest boundary of the controlled area shall not receive a dose greater than 5 rem to the whole body or any organ from any design-basis accident. The minimum distance from the spent fuel handling and storage facilities to the nearest boundary of the controlled area shall be 100 meters. [10 CFR 72.24(d), 72.24(m), 72.106(b), and 36(d)]
The DCSS must be designed to provide redundant sealing of confinement systems. [10 CFR 72.236(e)]
Storage confinement systems must have the capability for continuous monitoring in a manner such that the licensee will be able to determine when corrective action needs to be taken to maintain safe storage conditions. [10 CFR 72.122(h)(4) and 72.128(a)(1)]
The DCSS design must include inspections, instrumentation and/or control (I&C) systems to monitor the SSC that are important to safety over anticipated ranges for normal and off-normal operation. In addition, the applicant must identify those control systems that must remain operational under accident conditions. [10 CFR 72.122(i)]
When practicable, the DCSS must be designed on the basis of favorable geometry, permanently fixed neutron-absorbing materials (poisons), or both. Where solid neutron-absorbing materials are used, the design shall allow for positive means to verify their continued efficacy. [10 CFR 72.124(b)]
Storage systems must be designed to allow ready retrieval of spent fuel for further processing or disposal. [10 CFR 72.122(l)]
The DCSS must be designed to minimize the quantity of radioactive waste generated. [10 CFR 72.24(f) and 72.128(a)(5)]
The applicant must describe equipment and processes proposed to maintain control of radioactive effluents. [10 CFR 72.24(l)(2)]
To the extent practicable, the DCSS must be designed to facilitate decontamination. [10 CFR 72.236(I)]
The applicant must establish operational restrictions to meet the limits defined in 10 CFR Part 20 and to ensure that radioactive materials in effluents and direct radiation levels associated with ISFSI operations will remain as low as is reasonably achievable (ALARA). [10 CFR 72.24(e) and 72.104(b)]
SSC that are important to safety must be designed, fabricated, erected, tested, and maintained to quality standards commensurate with the importance to safety of the function to be performed. [10 CFR 72.24(c), 72.122(a), 72.122(f), and 72.128(a)(1)]
The DCSS must be designed for decommissioning. Provisions must be made to facilitate decontamination of structures and equipment and to minimize the quantity of radioactive wastes, contaminated equipment, and contaminated materials at the time the ISFSI is permanently decommissioned. [10 CFR 72.24(f), 72.130, and 72.236(I)]
The applicant must provide information concerning the proposed practices and procedures for decontaminating the site and facilities and for disposing of residual radioactive materials after all spent fuel has been removed. Such information must provide reasonable assurance that decontamination and decommissioning will adequately protect the health and safety of the public. [10 CFR 72.24(q) and 72.30(a)]
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The reviewer should verify that the applicant has provided either general or summary discussions of the SSC's design features, and both operational and accident conditions in a sufficiently clear manner that the applicant demonstrates a clear and defensible case that they have met the design criteria. In evaluating the principal design criteria related to DCSS SSC that are important to safety, reviewers should seek to ensure that the given design fulfills the following acceptance criteria:
In establishing normal and off-normal conditions applicable to the design criteria for DCSS designs, applicants should account for actual facility operating conditions. Design considerations should therefore reflect normal operational ranges, including any seasonal variations or effects.
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All members of the review team should review Section 2 of the SAR. Although RG 3.61 defines the standard format and content of an SAR, it does not address the different levels of detail expected in introducing component design criteria in SAR Section 2 and as compared with latter sections of the SAR. Consequently, reviewers for each section of the SAR should consider Section 2 in combination with additional details presented later in the SAR. In this SRP, evaluation of design criteria applicable to each of the relevant chapters of the SAR are discussed in detail in those chapters.
Inclusion of a separate section for design criteria in both the SAR and SER supports the staff's procedure of deliberately reviewing these criteria for acceptability apart from the proposed design and infrastructure of the system. This approach forms a "two-step" review process in which the acceptability of the detailed design criteria is separately stated. In-depth evaluation to assess satisfaction of these or other criteria is addressed in other sections of this SRP.
Although the design criteria presented in the SAR may be acceptable to the staff, the actual design may not meet either these criteria or the applicable regulatory requirements. It is also possible that the design criteria themselves, as presented in the SAR, may be unacceptable for application to a given DCSS design. As a result, the design may be unacceptable in that it does not meet the regulatory requirements, or the design may satisfy alternative criteria that are not described in the SAR, but are acceptable to the NRC staff. Reviewers should bring any of these situations to the immediate attention of NRC management.
Pay particular attention to instrumentation and other equipment (e.g., lifting devices and transport vehicles). In general, the NRC staff accepts that monitoring systems need not be classified as being important to safety. For example, a failure in the functioning of the pressure monitoring system does not directly result in a release of radionuclides. Additional justification for not considering such systems as being important to safety may be presented in later sections of the SAR and summarized in Section 2.
SSC designated as being important to safety should be included or referenced in the discussion of Design Features within the Technical Specifications provided in SAR Section 12.
Examine any limitations regarding the condition of the spent fuel. If damage that could be classified as a "Gross Cladding Defect" is allowed, the effects of such damage should be assessed in later sections of the SAR. If damaged rods have been removed from a fuel assembly, determine whether a need exists to replace them with dummy rods before loading into the cask. Note, the presence of an additional moderator will need to be addressed in the criticality analysis in SAR Section 6.
The release of fill and fission product gases from failed fuel rods increases the pressure in the cask cavity, as well as increasing the potential source-term in the event of confinement failure. Consequently, the applicant should provide information regarding the fill/fission product gas present in the fuel as well as the free volume in the cask cavity to enable reviewers to evaluate the pressure in the cask cavity resulting from cladding failure during storage. For the purpose of calculating internal cask pressures, the NRC staff has accepted the following bounding assumptions regarding the minimum percentages of fuel rods to have failed (and released their gases):
Pay particular attention to the specification of burnup, cooling time, and decay heat generation rate. These parameters are generally not independent, and the manner in which they are specified and combined can significantly affect the maximum allowed cladding temperature, as discussed in Chapter 4 of the SRP.
Note the specification of enrichment limits. As discussed in Chapter 5 of the SRP, the criticality evaluation is based on the highest enrichment (for a given fuel assembly), while the shielding source term, especially for neutrons, should be based on the lowest enrichment (for a given burnup).
The SAR will typically list various fuel assemblies that can be stored in the DCSS. In general, no one type of fuel assembly will be bounding for all analyses. Ensure that the applicant has justified which specifications are bounding for each of the evaluations presented in subsequent sections of the SAR. Specifications used in these analyses should also be clearly identified or referenced in Section 12 of both the SAR and SER.
If the applicant requests permission for the storage of non-fuel core components in the cask, review the relevant detailed specifications, conditions, and constraints presented in the SAR. These specifications should be at least as detailed as the applicable information presented for the fuel designs, to provide the reviewer with a basis for a determination that the relevant safety functions of the DCSS will be maintained.
At a minimum the NRC staff has generally addressed the conditions discussed below; however, other conditions may be relevant depending on specific details of the DCSS design. Reviewers should pay particular attention to special design features and how these might be affected both by other external conditions and other DCSS components.
"Normal" conditions (including conditions involving handling and transfer) and the extreme ranges of normal conditions are presumed to exist during design-basis accidents or design-basis natural phenomena, with the exception of irrational or readily avoidable combinations. For example, an earthquake or tornado may occur at any time and in combination with any "normal" condition. By contrast, it can be presumed that transfer, loading, and unloading operations would not be conducted during a flood.
"Off-normal" conditions and events are presumed to occur in combination with normal conditions that are not mutually exclusive. Nonetheless, it is not required that the SAR analyze or the system be designed for the simultaneous occurrence of independent off-normal conditions or events, design-basis accidents, or design-basis natural phenomena.
Conditions involving a "latent" equipment or instrument failure or malfunction (that is, one that occurs and remains undetected) should be presumed to exist concurrently with other off-normal or design-basis conditions and events.Typical latent malfunctions include a misreading instrument that is not detected as part of routine procedures; an undetected ventilation blockage; or undetected damage from an earlier design-basis event or condition if no provisions exist for detection, recovery, or remediation of such conditions.
For normal , off-normal and accident conditions, reviewers should verify that the applicant has defined appropriate operating and accident scenarios. For these scenarios the applicant should include in the SAR a comprehensive evaluation of the effects of such scenarios on the SSC important to safety. Applicant's evaluations should demonstrate that the requirements of 10 CFR Part 72 .106 as well as 10 CFR Part 20 have been met.
If appropriate, the following design bases should be included as operating controls and limits in Section 12 of both the SAR and SER:
(1) Normal Conditions
For a given spent fuel specification, the primary external conditions that affect DCSS performance are, the ambient temperatures, insolence, and the operational environment experienced by the DCSS.
The NRC accepts as the maximum and minimum "normal" temperatures the highest and lowest ambient temperatures recorded in each year, averaged over the years of record. For the SAR, the applicant may select any design-basis temperatures as long as the restrictions they impose are acceptable to both the applicant and the NRC. If the cask is also designed for transportation, the temperature requirements of 10 CFR Part 71(12) could determine the design basis temperatures for storage.
For storage casks, the NRC staff accepts a treatment of insolence similar to that prescribed in 10 CFR Part 71.71 for transportation casks. If the applicant selects another design approach, it must be justified in the SAR.
The operational environment experienced by the DCSS under normal conditions includes the manner in which the cask is loaded, unloaded, and lifted. Occupational dose rates will in part, depend on whether the cask is sealed in a wet or a dry environment. Fuel cladding temperatures may also be affected. The manner in which the cask is lifted will determine the load on the trunnions and/or lifting yoke. The orientation of the cask (vertical or horizontal) and its height above ground during transport to the ISFSI will establish initial conditions for the drop accidents discussed below.
(2) Off-Normal Conditions
SARs generally address several off-normal conditions. These should include variations in temperatures beyond normal, failure of 10 percent of the fuel rods combined with off-normal temperatures, failure of one of the confinement boundaries, partial blockage of air vents, human error, out-of-tolerance equipment performance, equipment failure, and instrumentation failure or faulty calibration.
(3) Accident Conditions
The staff has generally considered that the following accidents should be evaluated in the SAR. Because of the NRC's defense-in-depth approach, each should be evaluated regardless of whether it is highly unlikely or highly improbable. These do not constitute the only accidents that should be addressed if the SAR is to serve as a reference for accidents for the site-specific application. Others that may be derived from a hazard analysis could include accidents resulting from operational error, instrument failure, lightning, and other occurrences. Accident situations that are not credible because of design features or other reasons should be identified and justified in the SAR.
(a) Cask Drop
The SAR should identify the operating environment experienced by the cask, as well as the drop events (i.e., end, side, corner) that could result. Generally the design basis is established either in terms of the maximum height to which the cask may be lifted when handled outside the reactor site spent fuel building or in terms of the maximum acceleration that the cask could experience in a drop.
(b) Cask Tipover
Although cask system supporting structures may be identified and constructed as being important to safety (i.e. designed to preclude cask tipovers), the NRC considers that cask tipover events should be analyzed. In some cases, cask tipover may be determined to be a credible hazard, and the associated analysis should reflect the conditions (e.g., heights and accelerations) associated with that hazard.
In the absence of an identified hazard, the NRC has accepted a non-mechanistic cask tipover about a lower corner onto a receiving surface from a position of balance with no initial velocity. The receiving surface for a horizontal or vertical drop may be either an unyielding hard surface; or, the receiving surface may be modeled as a reinforced concrete pad on an engineered fill(13),(14). The NRC has also accepted analysis involving the dropping of a cask with its longitudinal axis in the horizontal position that, with analysis of a vertical axis drop, could bound a non-mechanistic tipover case.
(c) Fire
The fire conditions postulated in the SAR should provide an "envelope" for subsequent comparison with site-specific conditions. The NRC accepts the methods discussed in 10 CFR Part 71.73. The NRC staff also accepts that the applicant may consider a fire based upon the limited availability of flammable material at an ISFSI (e.g., only that associated with vehicles transporting or lifting the cask or possibly nearby foliage). Regardless of which approach the applicant takes, the SAR should specify and justify the bounding conditions for a "design basis" fire
(d) Fuel Rod Rupture
The regulations require that the cask be designed to withstand the effects of accident conditions and natural phenomena events without impairing its capability to perform safety functions. Consequently, the NRC has asserted and the applicant should assume, during the cask analysis for conditions resulting from design-basis accidents and natural phenomena, a release of 100 percent of the initial rod fill gases and a release of 30 percent of the fission product gases from the fuel rods into the cask interior. The remaining 70 percent of the fission product gases are presumed to be retained within the fuel pellet.
(e) Leakage of the Confinement Boundary
Casks are designed to provide the confinement safety function under all credible conditions. Nevertheless, the NRC staff considers that, for assessment purposes and to demonstrate the overall safety of the storage cask system, the DCSS should be evaluated for the effects of a confinement boundary failure. The SAR should identify this failure as a bounding release caused by a non-mechanistic event and the effects should be evaluated as described in the Sandia National Laboratories Report 80-2124(15).
(f) Explosive Overpressure
The conditions under which an ISFSI may be exposed to the effects of an explosion vary greatly among individual sites. Generally, explosive overpressure is postulated to originate from an industrial accident. The effects of various sabotage methods on cask systems were evaluated separately by the Division of Fuel Cycle Safety and Safeguards in developing appropriate regulations in 10 CFR Part 73(16). Therefore, explosive overpressures from sabotage events are not be considered in this SRP.
The extent to which explosive overpressure is addressed in the SAR directly affects the degree of site-specific review required. The principal concern in the SAR should be the effects of explosive overpressure on the storage system, rather than descriptions of hypothesized causes. Design parameters for blast or explosive overpressures should identify pressure levels as reflected ("side-on") overpressure, and should provide an assumed pulse length and shape. This discussion should provide sufficient information for licensees to determine if the effects of their site-specific hazards are bounded by the cask system design bases.
(g) Air Flow Blockage
For storage systems with internal air flow passages, the applicant should consider blockage of air inlets and outlets in an accident condition. The NRC staff considers that the effects of such an assumption should be utilized in determining the appropriate inspection intervals, and/or monitoring systems, for the DCSS.
(4) Natural Phenomena Events
The staff has generally considered that the following events should be evaluated in the SAR.
(a) Flood
The SAR should establish a design-basis flood condition. This condition may be determined on the basis of the presumption that the cask cannot tip over and the yield strength of the cask will not be exceeded. Alternatively, the SAR can show that credible flooding conditions have negligible impact on the cask design.
If the SAR establishes parameters for a design-basis flood, all of the potential effects of flood water and ravine flood byproducts should be recognized. Serious flood consequences can involve effects such as blockage of ventilation ports by water and silting of air passages. Other potential effects include scouring below foundations and severe temperature gradients resulting from rapid cooling from immersion.
(b) Tornado
The NRC staff accepts design-basis tornado wind loading as defined by RG 1.76 (Region 1)(17) and tornado missile impacts defined by NUREG-0800, Section 3.5.1.4(18). Design criteria should be established for the cask on the basis of these wind loading and missile impact definitions. The cask should not tip over and that the capability to perform the confinement safety function should not be impaired. The NRC considers that tornados and tornado missiles may occur without warning. The review should note that in general, the effects of a tornado missile bound those of a light general aviation aircraft directly impacting a DCSS.
(c) Earthquake
The SAR should state the parameters of the DBE. For ISFSIs at reactor sites, this is equivalent to the SSE used for analysis of nuclear facilities, under 10 CFR Part 50. An analysis for an "Operating-Basis Earthquake" (OBE) is not required for an DCSS SAR prepared in accordance with 10 CFR Part 72. Cask tipover accidents are analyzed, but tipover caused by an earthquake may not be a credible event.
(d) Burial under Debris
Debris resulting from natural phenomena or accidents that may affect cask system performance may be addressed in the SAR or may be left to the site-specific application. Such debris can result from floods, wind storms, or land slides. The principal effect is typically on thermal performance.
(e) Lightning
Lightning typically has a negligible effect on cask systems; however, the requirements of the Lightning Protection Code and National Electric Code should be applied to the design of the cask system structures. These codes should be cited as part of the general design criteria for the cask system (see Section II.3.a, above). Lightning should also be addressed as a natural phenomenon in the SAR if cask system performance may be affected if lightning affects a component that is important to safety.
(f) Other
10 CFR Part 72 identifies several other natural phenomena events (including seiche, tsunami, and hurricane) that should be addressed for spent fuel storage. The SAR may include these as design-basis events or show that their effects are bounded by other events. If they are not addressed in the SAR and they prove to be applicable to a specific site, a safety analysis is required prior to approval for use of the DCSS under either a site specific, or general license.
3. Design Criteria for Safety Protection Systems
Because RG 3.61 does not distinguish between the principal design criteria that should be presented in Section 2 of the SAR and those that should be deferred to subsequent sections, the applicant may take one of several approaches. SAR Section 2 may discuss these criteria in general terms (similar to the wording in Section II.3 above), with details provided in later sections. Alternatively, SAR Section 2 may present detailed discussions of selected (or all) criteria. Past applicants have generally selected the latter approach. Subsequent chapters of this SRP provide detailed discussions of the design criteria applicable to each functional area (e.g. structural, thermal) without regard to those that may have been presented in SAR Section 2.
Cask system components that are to be used in facility areas subject to review under 10 CFR Part 50 should satisfy both the requirements in 10 CFR Part 72 (with review guided by this SRP) and 10 CFR Part 50 (with review guided by NUREG-0800 and applicable portions of RG 3.53(19)). Acceptance of the cask system in areas covered by 10 CFR Part 50 license requirements is not addressed in this SRP for approval under 10 CFR Part 72. If a reviewer knows that the cask system will be used at a specific reactor site, the NRR project manager for that site should be so informed. The reviewer is reminded that a likely matter of interest to NRR is heavy loads.
Regardless of where the descriptions and associated criteria are located in the SAR, reviewers should include a summary description and evaluation of the safety protection systems in the "Design Criteria" section of the SER. The system descriptions should address the functions of the various system components in providing confinement, cooling, subcriticality, radiation protection of the public and workers, and spent fuel retrieval. Summary criteria for the performance of the system as a whole in providing for these capabilities or functions should also be described and evaluated. Reviewers should verify that the design-basis assumptions presented are consistent with and reasonable for actual site or facility conditions.
Criteria relating to redundancy, and allowable levels of response by the DCSS under normal, off-normal, and design-basis conditions and events should be described and evaluated. In general, no unacceptable degradation in physical condition or functional performance should result from normal or off-normal conditions. The design criteria regarding limits of permissible system response and degradation resulting from a DBE should be evaluated against the SSC capabilities to perform the principal safety functions. Considerations of permissible responses should include detectability and corrective actions that may be proposed as conditions of system use.
Table 2-1 summarizes design criteria (and design bases) that should generally be identified during the initial stages of the review. This listing may vary depending on the details of the cask design.
(a) Continuous Monitoring
The Office of the General Counsel (OGC) has developed an opinion as to what constitutes "continuous monitoring" as required in 10 CFR Part 72.122(h)(4). The staff, in accordance with that opinion has concluded that both routine surveillance programs and active instrumentation meets the intent of "continuous monitoring". Cask vendors may propose, as part of the SAR, either active i